Neutronic reactor



Sept. 2, H958 l... A. oHLlNGr-:R ETAL 2,350,447

NEUTRONIC REACTOR Filed Nov. 2, 1945 11 Sheets-Sheet 1 Sept. 2, 1958 L. A. OHLINGER ET AL 2,850,447

NEUTRONIC REACTOR Filed Nov. 2, 1945 11 sheets-sheet 2 FISE- l la/azesses:

NEUTRONIC REACTOR Filed Nov. 2. `1945 11 Sheets-Sheet 3 Sp. T1551 L. A. oHLlNGl-:R ETAL 21505447 NEUTRONIC REAc'roR Filed Nov. 2, 1945 11 sheets-sheet 4 l I 1 l oo ooo I ooooooo e,

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NEUTRONIC REAcToR Filed Nov. 2, 1945 ll Sheets-Sheet 5 L. A. oHLlNGl-:R ET AL 2,859,447

mauTRoNIcv REAcToR l1 Sheets-Sheet 6 Filed NOV. 2, 1945 Spt. 2, M95@ L. A. oHLxNGER E1' AL @8509447 NEUTRONIC REACTOR Filed Nov. 2, 1945 11 Sheets-Sheet 8 SPt- 2, 1958' L. A. OHLINGER Erm.' 2,850,447

NEUTRONIC REAcToR Filed Nov. 2, 1945 l1 Sheets-Sheet 9 Pt 2, 1958 L. A. oHLxNGER ET AL 2,850,447

NEUTRONIC I REACTOR Filed Nov. 2. 1945' 11 sheets-sheet 1o NEUTRNIC REACTOR Leo A. Ohlinger and Eugene P. Wigner, Chicago, Ill., Alvin M. Weinberg, Oak Ridge, Tenn., and Gale J. Young, Chicago, Ill., assignors to the United States of America as represented by the United States Atomic Energy Commission Application November 2, 1945, Serial No. 626,377

4 Claims. (Cl. 21M-193.2)

The present invention relates to the subject of neutronics, and more particularly to the charging of bodies containing iissionable material into and the discharging of same from a liquid cooled neutron chain reacting system also referred to as a neutronic reactor, or pile, the latter name having been originally adopted for the active portions of systems employing uranium or other fissionable bodies geometrically arranged in graphite or other moderator in the form of lattice structures. As a result of the chain reaction, when U238 is present (as in natural uranium), transuranic element 94239, known as plutonium, is produced. This material is lissionable and is valuable when added to natural uranium for use in a chain reacting system, as a fissionable body in lieu of or conjunction with natural uranium.

Natural uranium contains both uranium isotopes U235 and U238 in the ratio of l to 139. The U235 is the isotope lissionable by slow neutrons.

When iission occurs in the U235 isotope, the following reaction takes place:

where A represents light fission fragments having atomic masses ranging from 83 to 99 inclusive and atomic numbers from 34 to 45 inclusive; for example, Br, Kr, Rb, Zr, Y, Sr, Cb, Mo, Ma, Ru, and Rh; and B represents heavy fission fragments having atomic masses ranging from 127 to 141 inclusive and atomic numbers from 51 to 60 inclusive; for example, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, and Nd.

The elements resulting from the tssions are unstable and radio-active, with half-lives varying in length in accordance with the element formed.

The absorption of thermal or resonance neutrons by the U238 isotope give rise to the conversion of U238 to U239 which ultimately decays to transuranic element 94239. The reaction is as follows:

92238-1-11 -e 92239 [plus 6 m. e. v. of 'y rays,

not necessarily all of one frequency] 2.3 da s 93239 la 942394-13160() kv. upper energy limit. Also 2 'y rays, 400 kw., and 270 kv., about l of which are converted to electrons] Most of the neutrons arising from the fission process are set free with the very high energy of about one million electron volts average and are therefore not in condition to be utilized efliciently to create new thermal neu- Patented Sept. 2, 1958 tron issions in a ssionable body such as U235 when it is mixed with a considerable quantity of U238, particularly as in the case of natural uranium. The energies of the fission-released neutrons are so high that most of the latter would tend to be absorbed by the U238 nuclei, and yet the energies are not generally high enough for production of fission by more than a small fraction of the neutrons so absorbed. For neutrons of thermal energies, however, the absorption cross-section of U235, to produce fission, is a great deal more than the simple capture cross-section of Um; so that under the stated circumstances the fast ission neutrons, after they are created, must be slowed down to thermal energies before they are most effective to produce fresh fission by reaction with additional U235 atoms. If a system can be made in which neutrons are slowed down without excessive absorption until they reach low or thermal energies and then mostly enter into uranium rather than into any other element, a self-sustaining nuclear chain reaction can be obtained, even with natural uranium. Light elements, such as deuterium, beryllium, oxygen or carbon, the latter in the form of graphite, can be used as slowing agents. A special advantage of the use of the light elements mentioned for slowing down fast fission neutrons is that fewer collisions are required for slowing than is the case with heavier elements, and furthermore, the above-enumerated elements have very small neutron capture probabilities, even for thermal neutrons. Hydrogen would be most advantageous were it not for the fact that there may be a relatively high probability of neutron capture by the hydrogen nucleus. Carbon n the form of graphite is a relatively inexpensive, practical, and readily available agent for slowing fast neutro-ns to thermal energies. Recently, beryllium has been made available in sufficiently large quantities for test as to suitability for use as a neutron slowing material in a system of the type to be described. It has been found to be in every way as satisfactory as carbon. Deuterium while more expensive is especially valuable because of its low absorption of neutrons and its compounds such as dueterium oxide have been used with very eiective results.

However, in order for the premise to be fulfilled that the fast ssion neutrons be slowed to low or thermal energies in a slowing medium without too large an absorption in the U238 isotope of the uranium, certain types of physical structure should be utilized for the most efficient reproduction of neutrons, since unless precautions are taken to reduce various neutron losses and thus to conserve neutrons for the chain reaction the rate of neutron reproduction may be lowered and in certain. cases lowered to a degree such that a self-sustaining system is not attained.

The ratio of the number of fast neutrons produced by the iissions, to the original number of fast neutrons creating the issions, in a system of infinite size using specific materials is called the reproduction or multiplication factor of the system and is denoted by the symbol K. If`

K can be made suiiiciently greater than unity to create a net gain in neutrons and the system made suiiiciently large so that this gain is not entirely lost by leakage from the exterior surface of the system, then a self-sustaining chan reacting system can be built to produce power by nuclear lission of natural uranium. The neutron reproduction ratio r in a system of finite size differs from K by the leakage factor, and must be sufliciently greater than unity to permit the neutron density to rise exponentially. Such a rise will continue indefinitey if not controlled at a desired density corresponding to a desired power output.

During the interchange of neutrons in a system comprising bodies of uranium of any size in a slowing medium, neutrons may be lost in four ways, by absorption in the uranium metal or compound without producing fission, by absorption in the slowing down material, by absorption in impurities present in the system, and by leakage from the system. These losses will be considere/.l in the order mentioned.

Natural uranium, particularly by reason of its U238 content, has an especially strong absorbing power for neutrons when they have been slowed down to moderate energies. The absorption in uranium at these energies is termed the uranium resonance absorption or capture. It is caused by the isotope U238 and does not result in fission but creates the isotope U239 which by two successive beta emissions forms the relatively stable nucleus 94239. It is not to be confused with absorption or capture of neutrons by impurities, referred to later. Neutron resonance absorption in uranium may take place either on the surface of the uranium bodies, in which case the absorption is known as surface resonance absorption, or it may take place further in the interior of the uranium body, in which case the absorption is known as volume resonance absorption. It will be appreciated that this classification of resonance absorptions is merely a convenient characterizatio-n of observed phenomena, and arises, not because the neutron absorbing power of a U238 nucleus is any greater when the nucleus is at the surface of a body of metallic, or combined uranium, but because the absorbing power of U238 nuclei for neutrons of certain particular energies is inherently so high that practically all neutrons that already happen to have those energies, called resonance energies as explained above, are absorbed almost immediately upon their arrival in the body of uranium metal and uranium compound and thus in effect are absorbed at the surface of such body. Volume resonance absorption is due to the fact that some neutrons make collisions inside the uranium body and may thus arrive at resonance energies therein. After successfully reaching thermal velocities, about 40 percent of the neutrons are also subject to capture by U238 without fission, to produce U239 and eventually 94239.

It is possible, by proper physical arrangement of the materials, to reduce substantially uranium resonance absorption. By the use of light elements as described above for slowing materials, a relatively large increment of energy loss is achieved in each collision and therefore fewer collisions are required to slow the neutrons to thermal energies, thus decreasing the probability of a neutron being at a resonance energy as it enters a uranium atom. During the slowing process, however, neutrons are diffusing through the slowing medium over random paths and distances so that the uranium is not only exposed to thermal neutrons but also to neutrons of energies varying between the emission energy of ssion and thermal energy. Neutrons at uranium resonance energies will, if they enter uranium at these energies, be absorbed on the surface of a uranium body whatever its size, giving rise to surface absorption. Any substantial reduction of overall surface of the same amount of uranium relative to the amount of slowing material (i. e. the amount of slowing medium remaining unchanged) will reduce surface absorption, and any such reduction in surface absorption will release neutrons to enter directly in the chain reaction7 i. e., will increase the number of neutrons available for further slowing and thus for reaction with U235 to produce ssion.

For a given ratio of slowing material to uranium, surface resonance absorption losses of neutrons in the uranium can be reduced by a large factor from the losses occurring in a mixture of tine uranium particles and a slowing medium, if the uranium is aggregated into substantial masses in which the mean radius of the aggregates is at least 0.25 centimeter for natural uranium metal and when the mean spatial radius of the bodies is at least 0.75 centimeter for the oxide of natural uranium (U02). Proportionate minimums exist of other uranium compounds the exact minimum value being dependent upon the uranium content and the density of the product. An important gain is thus made in the number of neutrons made directly available for the chain reaction. A similar gain is made when the uranium has more than the natural content of tissionable material. Consequently, where a maximum K factor is to be desired we place the uranium in the system in the form of spaced uranium masses or bodies of substantial size, preferably either of metal, oxide. carbide, or other compound or combinations thereof. The uranium bodies can be in the form of layers, rods or cylinders, cubes or spheres, or approximate shapes, dispersed throughout the graphite, preferably in some geometric pattern. The term geometric is used to mean any pattern or arrangement wherein the uranium bodies are distributed in the graphite or other moderator with at least either a roughly uniform spacing or with a roughly systematic non-uniform spacing, and are at least roughly uniform in size and shape or are systematic in variations of size or shape to produce a volume pattern conforming to a roughly symmetrical sfstem. lf the pattern is a repeating or rather exactly regular one, a system embodying it may be Conveniently described as a lattice structure.

The number of neutrons made directly available to the chain reaction by aggregating the uranium into separate bodies spaced through the slowing medium is a critical factor in obtaining a self-sustaining chain reaction utilizing natural uranium and graphite. The K factor of a mixture of tine uranium particles in graphite, assuming both of them to be theoretically pure. would only be about 0.785. Actual K factors as high as 1.07 have been obtained using aggregation of natural uranium in the bes-t known geometry. and with as pure materials as it is presently possible to obtain.

Assuming theoretically pure carbon and theoretically pure natural uranium metal, both of the highest obtainable densities, the maximum possible K factor theoretically obtainable is about 1.1 when the uranium is aggregated with optimum geometry. Still higher K factors can be obtained by the use of aggregation in the case of uranium having more than the naturally occurring content of fissionable elements. Adding such ssionable material is termed enrichment of the uranium.

It is thus clearly apparent that the aggregation of the uranium into masses separated in the slowing material is one of the most important factors entering into the successful construction of a self-sustaining chain reacting system utilizing relatively pure natural uranium in a slowing material such as graphite in the best geometry at present known, and is also important in obtaining high K factors when enrichment of the uranium is used.

Somewhat higher K factors are obtainable where moderators such as deuterium oxide or beryllium are used. Thus with beryllium it is possible to secure a K factor as high as 1.18 and with deuterium oxide K factors as high as 1.3 may be obtained.

The thermal neutrons are also subject to capture by the slowing material. While carbon and beryllium have very small capture Cross-sections for thermal neutrons, and deuterium still smaller, an appreciable fraction of thermal neutrons (about 10 percent of the neutrons present in the system under best conditions with graphite) is lost by capture in the slowing material during diffusion therethrough. It is therefore desirable to have the neutrons reaching thermal energy promptly enter uranium.

In addition to the above-mentioned losses, which are inherently a part of the nuclear chain reaction process, impurities present in both the slowing material and the uranium add a very important neutron loss factor in the chain. The effectiveness of various elements as neutron absorbers varies tremendously. Certain isotopes of elements such as xenon, boron, cadmium, samarium, gadolinium, and some others, if present even in a few parts per million, could prevent a sulf-sustaining chain reaction from taking place. It is highly important, therefore, to remove as far as possible all impurities capturing neutrons to the detriment of the chain reaction from both the slowing material and the uranium. If these impurities, solid, liquid, or gaseous, and in elemental or combined form, are present in too great quantity, in the uranium bodies or the slowing material or in, or by absorption from, the free spaces of the system, the selfsustaining chain reaction cannot be attained. The amounts Iof impurities that may be permitted in a system, vary with a number of factors, such as the specific geometry of the system, and the form in which the uranium is used-that is, whether natural or enriched, whether as metal or oxide-fand also factors such as the Weight ratios `between the uranium and the slowing down material, and the type of slowing down or moderating material userl-for example, whether deuterium, graphite or beryllium.

The strong absorbing action of some elements renders a self-sustaining chain reacting system capable of control. By introducing neutron absorbing elements in the form of rods or sheets into the interior of the system, for instance in the slowing material between the uranium masses, the neutron reproduction ratio r of the system can be changed in accordance with the amount of absorbing material exposed to the neutrons in the system. A sucient mass of the absorbing material can readily be inserted into the system to reduce the reproduction ratio of the system to less than unity and thus stop the reaction.

When the uranium and the slowing material are of such purity and the uranium is so aggregated that fewer neutrons are parasitically absorbed than are gained by iission, the uranium will support a chain reaction producing an exponential rise in neutron density if the overall size of the system is sufciently large to overcome the loss of neutrons escaping from the system.

Thus overall size of the system is very important and it will vary, depending upon the K factor of the system, and upon other things. If the reproduction factor K is greater than unity, the number of neutrons present will increase exponentially and indefinitely, provided the structure is made sufficiently large. If, on the contrary, the structure is small, with a large surface-to-volume ratio, there will be a rate of loss of neutrons from the structure by leakage through the outer surfaces, which may overbalance the rate of neutron production inside the structure so that a chain reaction will not be selfsustaining For each Value of the reproduction factor K greater than unity, there is thus a minimum overall size of a given structure known as the critical size, above which the rate of loss of neutrons by diffusion to the walls ofthe structure and leakage away from the structure is less than the rate of production of neutrons within the system, thus making the chain reaction self-sustaining. The rate of diffusion of neutrons away from a large structure in which they are being created through the exterior surface thereof may be treated by mathematical analysis when the value of K and certain other constants are known, as the ratio of the exterior surface to the volume becomes less as the structure is enlarged.

In the case of a spherical structure employing uranium bodies imbedded in graphite in the geometries disclosed herein and without an external reflector the following formula gives the critical overall radius (R) in feet:

where `C is a constant that varies slightly with geometry of the lattice and for normal graphite lattices may have a value close to 7.4.

For a rectangular parallelepiped structure rather than spherical, the critical size can be computed from the formula where a, b, and c are the lengths of the sides in feet. The critical size for a cylindrical structure is given by the formula, irrespective of the shape of the uranium bodies: cylinder height h ft., radius R ft.

tion at rated power, have a reproduction ratio (r) up toI about 1.005 at operating temperature with all control absorbers removed are easily controlled. The size at which this reproduction ration can be obtained may be computed from modications of the above formulae for critical size. For example, for spherical active structures the formula C K-T may be used to nd R when K is known and r is somewhat over unity. The same formula will, of course, give r for given structures for which K and R are known.

Critical size may be attained with a somewhat smaller structure by utilizing a neutron reflecting medium surrounding the surface of the active structure. Forexample, a 2 foot thickness of graphite having low impurity content, completely surrounding a spherical structure is effective in reducing the diameter of the uranium bearing portion by as much as 2 feet.

The rate of production of element 94239 will depend on the rate of neutron absorption by U238 and is also proportional to the rate at which ssions occurs in U235. This in turn is controlled by the thermal neutron density existing in the reactor While operating. Thus for maximum production of element 94239, it is essential that the thermal neutron density be at a maximum value commensurate with thermal equilibrium.

Considerable heat is generated during a neutronic reaction primarily as the result of' the fission process. Following are tables showing more specifically the type of heat generated in the reactor.

SUMMARY BY TYPE SUMMARY BY LOCALE WHERE HEAT IS GENERATED M. e. v./ Percent fission In uranium... 184 92 In moderator-. 12 6 Outside pile 4 2 SUlVIB/IARY BY TYPE AND LOCALE When the system is operated for an extended period of time at a high production output of element 94239, the large amount of heat thus generated must be removed in order to stabilize the chain reaction. Most of the heat in an operating device is generated as the result of the nuclear ssions taking place in the U235 isotope. Thus, the rate of heat generation is largely proportional to the rate at which the ssions take place. In other words, if the rate of generation of neutrons is increased, a greater amount of coolant must be passed through the reactor in order to remove the heat thus generated to avoid damage, particularly at the central portion of the pile, by excessive heat. Thus, the highest obtainable neutron density at which a system can be operated for an extended period of time is limited by the rate at Which the generated heat can be removed. That is to say, the maximum power output of a system is limited by the capacity of the cooling system. An effective cooling system is therefore a primary requirement for high power operation of a neutronic reactor and it has been found that this cooling may be accomplished most effectively by passage of the coolant in contact with or in close proximity to the uranium.

After the neutronic system has operated for a period of time suicient to cause a quantity of element 94239 to be produced, it may be desirable to remove at least some of the uranium rods from the reactor in order to extract element 94239 and the radioactive fission products, both being formed in the uranium rods or for other purposes. The present invention relates more particularly to the removal of uranium bodies from the neutronic reactor.

ln many neutronic reactors a neutron density variation occurs across the reactor; that is, the neutron concentration at the periphery s relatively small and in creases to a maximum value at the center. Actually, therefore, since the rate of production of element 94239 is dependent upon the neutron density, the reactor will have Zones which may be likened to three dimensional shells, the average concentration of element 94239 being uniform throughout any given zone. ln a reactor built in the form of a sphere these would, of course, be in the shape of concentric spheres of different diameters, while one built in the shape of a cylinder would have similar zones but of different shapes,

Where this varation in concentration exists in a reactor it is often desirable to resort to a systematic schedule of removal depending upon the time of operation and the location of the uranium for removing and discharging uranium metal that has been subjected to neutron bombardment. In the case of a new system of this character the operation would normally continue until the metal in the center portion of the reactor reaches a desired content of element 94239 at which time this metal would be removed and replaced with fresh metal. The next removal then would be from the section next adjacent to the center section of the reactor where the desired content of element 94239 is reached after further operation. The process would then proceed with the removal of the metal at Various times until the metal recharged at the center of the reactor has reached the desired content of element 94239. This would then be replaced and the process of progressing towards the periphery continued with periodic return to more central areas. Since the neutron density in the central areas of such a reactor would, ordinarily, greatly exceed the neutron density near the periphery, the metal in the central areas may be replaced several times for each replacement of the metal near the periphery. A removal schedule can be developed by calculation and checked by actual experience after the system has been placed in operation.

Different schedules may be developed with other reactors having dierent .reactivity curves. For example, certain reactors are constructed in a manner such that the neutron concentration is substantially uniform throughout a large volume of the reactor. ln such a case the schedule for removal of uranium bodies may be modified accordingly.

Since the heat generated in the reactor results from issions in the uranium, it is evident that this heat is not formed uniformly throughout the reactor but that it must vary across the reactor with the local rate at which issions occur and element 94239 formed. Consequently, the relative values for the production of ele,- ment 94239 apply also to heat distribution; that is, the heat generated may increase from a minimum at the outer surface of the reactor to a maximum at the center in certain reactors.

As the total weight of the radioactive fission elements is proportional to that of the 94239 at the time of issions, it might be assumed that the amounts of these radioactive fission elements and of 94239 present in metal removed from the reactor are also of the same proportion. This is not true, however, as the fiassion elements when produced are highly radioactive and immediately start to decay, some with short half-lives and others with longer halfelives until, through loss of energy, these unstable iission elements arrive at a stable non-radioactive element or isotope and no longer change. The 94239 on the other hand is a relatively stable element when formed, having a radioactive half-life of about 104 years.

At the start of the reaction in new metal the radioactive ssion elements and the 94239 both increase in amounts. After a certain period of operation during which time the metal is subjected to intense neutron bombardment the radioactive fission elements will reach a state of equilibrium and from that time on the amounts of these radioactive elements remain constant, as the fission elements with shorter half-lives are reaching a stable condition at the same time new ones are being produced. The amount of the stable end products of ssion, however, continues to increase with the increase in elements 94239. Consequently, the rate of formation of the iission end products is dependent upon the location of any particular metal in the reactor, and the power at which the system operates controls the maximum radioactive fission element content regardless of the length of time the system operates after equilibrium occurs. The quantity of element 94239 on the other hand, and of the iinal and stable end products of fission continue to increase as the operation of the system continues. The amounts of both 94239 and ssion end products present are controlled only by the location of the metal in the reactor and the time and power of operation. The highly radioactive iission elements may, therefore, vary from a substantial percentage of the weight of element 94239 preesnt in the metal at the center of the reactor after a short period of operation, to a very small percentage in metal from a position near the periphery of the reactor after an extended operating period at a given power.

It is not to be assumed, however, that the fact that equilibrium can be obtained between the original highly radioactive ssion elements and the stable fission end products that all radioactivity will cease when the original fission elements have been permitted to decay for a time equal to the equilibrium period. For example, many of the original fission elements have long half-lives that, taken together with their successive radioactive disintegration products existing long after the fission elements having a shorter half-life have decayed, renders the uranium still radioactive, especially after prolonged bombardment at high neutron densities. In addition, the successive radioactive disintegration products of the original shorter lived fission elements may still be preesnt.

The equilibrium radioactivity is so intense that metal taken from the reactor for the recovery of element 94239 and fission products immediately after bombardment at high neutron densities will heat spontaneously due to self absorption of the intense radioactivity of the remaining radioactive fission products. The amount of heat given off as the result of the spontaneous heating will depend particularly on three factors: (l) the concentration of element 94239 and fission products in the metal; (2) the period of time for continuous operation required to reach this concentration; and (3) the elapsed time since the reactor was shut down and the metal was removed.

The metal from the center of the reactor in a system operating at a high power output, for example, at a 94239 concentration of l to 2,000, if not cooled, can increase in temperature at the rate of about 2000 C. per hour one day after the neutron activity of the system has been shut down. After 3() days shut down following an operation y of 100 days at an output of 500,000 kilowatts, the average temperature rise can be approximately 572 C. per hour. The uranium metal of the type used in the chain reacting systems herein under consideration melts at about 1100D C.

Under these conditions uranium bombarded with neutrons for an extended period of time at high rates of power output can be safely removed from the reactor under one of the following methods:

(l) The neutron activity of the system is shut down and the uranium is kept in the reactor and continuously cooled until the radioactivity decays to a point where the metal can be removed without melting in ambient air. This procedure may require that the metal remain in the reactor for a period of from 30 to 50 days after the neutron bombardment has ceased.

(2) The neutron activity of the system is shut down and the uranium is kept in the reactor with the cooling system in operation for only a few days to permit the most violent radioactivity to subside and then the metal is removed from the reactor with the cooling discontinued during the removal except for cooling by the atmosphere or by water spray. The metal is then promptly placed under more efficient cooling conditions before the temperature of the uranium has become excessive.

(3) The neutron activity of the system is shut down and the uranium is removed while cooling the uranium body at least to an extent sufficient to prevent the temperature from becoming excessive. This modification of the present invention is particularly effective.

It is also important, of course, from the point of view of biological safety of operating personnel that adequate shielding be provided to absorb the strong gamma radiations from the fission products present in the active uranium while being removed from the reactor. The neutron activity in the reactor completely ceases within 30 minutes after shut down of the neutronic reaction during which period delayed neutrons are being emitted from fission fragments. In no case then should the uranium be removed from the reactor immediately following shut down of the neutronic reaction, but sufficient time should be given to permit all delayed neutrons to be emitted. Thus, the shielding required during the removal of the uranium rods from the system is primarily intended to protect personnel from gamma radiations. As stated above, immediately following shut down of the neutronic reaction, there are many short lived radioactive fission elements in the uranium causing the gamma radiation to l0 be very intense. Many of these elements decay into more stable products within the rst thirty minutes following shut down of the reaction. Thus, the fission products lose a large amount of their radioactivity during this period.

While the method of extracting the fission products and element 94239 from the bombarded uranium taken from the reactor forms no part of the present invention, the fission products and element 94229 are removable and when removed are extremely useful. The radioactive fission products are valuable for use as radiation sources, many having long half lives with high energy gamma radiation sufficient for radiography of even heavy metal castings. In addition, some of the fission products are useful as radioactive tracers in biological and physiological research and are in demand by the medical profession.

Element 94239 is exceptionally useful because it is tissionable by slow neutrons in the same manner as the uranium isotope 92235 contained in natural uranium. The separation of 92235 from 92238 in natural uranium is-extremely difficult since both are isotopes of the same element and these isotopes vary only a small percentage in comparative weight. Element 941239 on the other hand, is a different element from uranium, having different chemical properties than uranium, and therefore can be chemically separated from uranium. After separation, for example, element 94239 can be added to natural uranium to supplement the 94235 content, thus increasing the amount of fissionable material in the uranium. This enriched uranium can then be used in neutronic systems making it possible to provide more cooling facilities, for example, than can be used in a system of the same geometry employing only natural uranium. Thus, an enriched neutronic system may provide a greater power output than would be possible in a natural uranium system having the same geometry.

To summarize the present invention is concerned with the unloading and loading of a liquid cooling neutronic reactor particularly after it has operated for a substantial period of time. Uranium or similar fissionable material which has been subjected to neutron bombardment is highly radioactive. The present invention provides means for removing such highly radioactive material without hazard to personnel. If the uranium or other fissionable material has been operated under conditions such that substantial heat has been evolvedthe problem of removing the product is made more ditlicult due to the self heating phenomenon which occurs after neutrom'c reaction has been discontinued. We found that this difficulty may be overcome to a substantial degree by cooling the iissionable composition during its removal, and thereafter cooling this composition immediately after its removal from the reactor until this self heating has been substantially reduced. Further we have found that the degree and period of cooling required in each case may be decreased by cooling the fissionable composition in the reactor after the reactor has been shut down and the self-sustaining chain reaction discontinued. if such preoooling is continued for several days cooling during removal may be dispensed with and if the precooling is continued for a long period of time, for example, 30 days or more cooling after removal may not be required. Apparatus suitable for performance of the above process has been provided in accordance with this invention. Particularly valuable modifications have been provided to permit removal of the lissionable bodies without 4shutting down the reactor for long periods of time.

The foregoing constitute some of the principal objects and advantages of the present invention, others of which will become apparent from the following description, read in conjunctio-n with the drawings, in which:

Fig. 1 is a schematic view of the reactor with its external cooling system;

Fig. 2 is a more or less diagrammatic, cross-sectional view, partly in elevation of the neutronic reactor and the loading and unloading means therefore;

Fig. 3 is a plan view of the exterior of the reactor and the loading and unloading means;

Fig. 4 is an elevation view of one-half of the inlet side of the reactor disclosing the lead shield, the elevator structure being removed;

Fig. 5 is a cross-sectional view, partly in elevation, taken through the center of the reactor;

Fig. 6 is an enlarged fragmentary View of a portion of the reactor, showing the intake end of the reactor and details of the uranium containing tubes and the associated valves and piping;

Fig. 7 is a view similar to Fig. 6 taken at the outlet end of the reactor;l

Fig. 8 is an enlarged, detailed View disclosing the loading valves and portions of the loading and unloading cars;

Fig. 9 is an enlarged, sectional view, partially in elevation, disclosing the details of the water inlet manifold at the end of one tube;

Fig. 10 is a transverse sectional view through a loaded tube, the view being taken on the line 10-10 of Fig. 9;

Fig. 11 is a plain view of the loading car;

Fig. 12 is a longitudinal sectional view, partially in elevation, and taken on the line 12--12 of Fig. 11;

Fig. 13 is a front elevation view of the car and looking in the direction 13-13 of Fig. 11;

Fig. 14 is a transverse sectional view through the car and taken on the line 14--14 of Fig. 15;

Fig. l5 is a side elevational View looking in the direction of arrows 15-15 of Fig. 14;

Fig. 16 is a longitudinal sectional view, partially in elevation of the unloading car;

Fig. 17 is a transverse sectional view taken on the line 17-17 of Fig. 16; and

Fig. 18 is a longitudinal sectional view through a uranium rod and a portion of an adjoining rod, part of the lirst rod being broken away.

GENERAL ORGANIZATION An entire neutronic reactor system accomplishing the objects hereinbefore set forth comprises broadly a power unit including a neutronic reactor, a complete heat extracting or cooling circuit adapted to remove heat from the reactor, an effective control system for regulating the operation of the neutronic reactor, and provisions for loading and discharging said bodies from the reactor after they have been subjected to neutron bombardment for a predetermined period of time. We prefer to illustrate our invention in conjunction with a reactor employing uranium rods disposed in a moderator of graphite. Referring to Figs. 1-5, inclusive, the reactor is generally indicated at 25 and comprises broadly a graphite moderator 26 and a plurality of horizontally disposed passages lined with aluminum tubes 27 extending throughout the width of the graphite. Short uranium rods 120 are placed in the aluminum tubes 27 as will be brought out hereinafter and these rods are adapted to be charged into the reactor from a loading car 30 and discharged from the opposite side of the reactor into an unloading car, generally indicated at 31. A control rod 179 and a safety rod 180 are shown diagrammatically. Their functions will be described hereinafter.

The cooling circuit Irrespective of the form the reactor may take, the heat may be extracted by circulating a coolant through the reactor in heat exchange relationship with the uranium rods. If the power unit is located near a relatively pure body of water such as a river of sutiicient size to supply the necessary quantity of water to extract the required amount of heat, then this river water can be passed through the reactor after first being processed to obtain the required purity. Under these conditions the water after passing through the reactor is returned to the river.

In the event a natural body of water is not available, then a supply of coolant must be provided. In this event, it is desirable to recirculate the coolant through the reactor many times and thus reduce to a minimum the total quantity required. The coolant leaving the reactor may carry with it certain free gases such as hydrogen and oxygen if water is used, which should be eliminated from the cooling circuit. For this purpose, the coolant leaving the reactor is passed through a ilash tank where these gases are removed. The coolant is then cooled and inally pumped back through the reactor and recirculated.

For purposes of illustration an external circulating system for the coolant is shown in Fig. l, wherein water, as the coolant, after passing through the reactor 25 to extract heat, is treated and filtered and then passed in heat exchange relationship with a secondary coolant and {inally is recirculated through the reactor.

In Fig. l, the reactor is diagrammatically shown at 25, the water inlet header being illustrated at 40. The cooling water is discharged from the reactor into an outlet header 20 and then enters the pipe 41 through which it is conveyed to a ilash tank 42, after passing through a throttling valve 41a. The water leaves the ash tank 42 through pipe 43, passes through heat exchanger 44 where it is cooled by flowing in heat exchange relationship with a cooler liquid, and then flows through pipe 45 and is returned to the reactor 25 by the pump 46 through pipe 47.

The water entering the heat exchanger 44 has a temperature only slightly less than the boiling point at the existing pressure. The water leaving the heat exchanger 44 has a temperature of about 95 F., this cooling being accomplished by transferring the heat to the cooler fluid in a secondary cooling system.

This secondary system includes a cooling tower generally indicated at 54, a pump 55, the heat exchanger 44 and suitable piping 56. The cooling fluid in the cooling tower 54, and for purposes of illustration water has been selected for this fluid, is collected in a reservoir 57 at the bottom of the tower from which the water is withdrawn by pump 55 and passed to the heat exchanger 44. This water entering the heat exchanger 44 is at a temperature of about F. and leaves the heat exchanger at a temperature of about F. This hot water passes through pipe 56 into spray head 58 disposed adjacent to the top of the cooling tower 54. The hot water is sprayed in a tine mist into the cooling tower 54, and mixes with air circulated through the cooling tower by a blower 59. Evaporation takes place resulting in the cooling of the water in the cooling tower 54 so that the water collected in the reservoir 57 has been effectively cooled by this process of evaporation. Losses due to vaporization may be replaced as will be understood in the art. Louvered openings 60 are provided in a wall of the cooling tower 54 through which air is discharged.

A portion of the cold water leaving pump 5S is bypassed from the secondary cooling system through a pipe 62 feeding condenser coils 52 and is returned to the secondary cooling system through pipe 63.

For purposes of illustration, distilled water is used in both cooling circuits and conventional means (not shown) may be provided for replenishing water lost by evaporation or leakage. Inhibitors may be added to the water to reduce the corrosive effect on the metal in the system. A mixture of potassium or ammonium phosphate and silicate is satisfactory for this purpose.

The entire primary water circulating system is provided with radiation shielding to protect personnel from the harmful effects of gamma radiations. As shown, the ash tank 42 is surrounded by a cylindrical concrete wall 64 extending above the top of the flash tank 42 and forming an enclosure which is completely lled with water. A similar wall 65 surrounds the heat exchanger 44 and the tank t formed thereby likewise is filled with water. piping is shown disposed underground.

For purposes of illustration only a diagrammatic showing is made at 66 of a shield surrounding the reactor 25. More complete details of this latter shield are brought out elsewhere. Similarly a diagrammatic showing of a shield surrounding the pump 46 is illustrated at 67.

All of the The power unit Since energy is extracted in the form of heat, the source of heat may be termed the power unit. Referring to Figs. 25, inclusive, the power unit including reactor 25, is shown. The reactor 25 is mounted on a concrete base '70 which rests on the ground 71. Vertical concrete walls 72, 73, 74 and 75 are disposed on the four sides of the reactor 25 extending upwardly from the base 70 to a position substantially above said reactor. Referring particularly to Figs. 4 and 5, the reactor 25 comprises a moderator 26 of graphite blocks in the form of a horizontally disposed cylinder through which are arranged a plurality of horizontal tubes 27, this entire assembly, except for the front and back faces, being surrounded by a graphite reilector 77 of about 50 centimeters in thickness.

The graphite moderator 26 may be built of graphite blocks cut to a size convenient for handling. The aluminurntubes 27 are inserted in holes drilled through the graphite blocks that are aligned to form continuous channels or passages through the length of the moderator 26. The entire active portion of the reactor 25 is enclosed in a gas-tight shield comprising two steel end sheets 79 and 80, a steel side sheet 81 and a bottom sheet 81a, welded to forni a gas-tight enclosure.

Spanning the distance between the concrete walls 72 and 74 are three intermediate spacing beams 84, 85 and 06.

Adjoining the steel end sheet 80 and extending parallel thereto is a partitioned tank that is filled with steel shot and water forming shield 76. A similar shield 276 is positioned on the opposite end of the moderator adjacent to end sheet 79. These shields will be described more fully hereinafter.

As shown in Figs. 2, 3, and 5, a space is provided between the end walls 73, 75 and the side sheet 81. Water lls this space and covers the top of the reactor forming a water shield 93. Thus except for the front and rear faces the entire active portion of the reactor is immersed in water.

Reactor 25 is positioned on a plurality of spaced stringers 137 in the form of I-beams resting on concrete. Between the stringers 137 a plurality of passages are provided through which water is circulated by a motor driven propeller 138. As shown in Fig. the water level is maintained at a pre-determined level by means of a lioat controlled valve 139 that is connected to a water supply.

Referring to Fig. 6 for the intake end and Fig. 7, for the outlet end the arrangement of aluminum tube 27 and its supports inserted in the graphite moderator 26 is shown. The moderator 26 is broken away and only a small portion is shown, but aluminum tubes 27 extend throughout the entire length of the cylindrically shaped graphite moderator 26 and as shown a continuous passage is provided for the uranium rods from the inlet face of the reactor through iron and water shields, the graphite moderator, the outlet-side iron and water shield and nally into the unloading car. End tube 27 terminates inside of a water distribution head 100 on the intake side of the reactor and in a similar distribution head 101 on the outlet side. A steel sleeve 102 extending from the water distribution head 100 to the graphite moderator 26 surrounds each aluminum tube 27 to support the portion of the aluminum tube that is unsupported by the graphite. A similar steel sleeve 103 extends from the moderator 26 to distribution head 101. Sleeves 102 and 103 terminate in bushings 104 set in the graphite. The sleeves 102 and 103 pass through shields 76 and 276 inside of steel tubes 105 that are welded to the shield tanks to form watertight passages for said sleeves 102 and 103. Steel tube 105 has inside taper to allow for differential horizontal and vertical expansion of the pile and shields. A steel shield collar 106 surrounds each sleeve 102 and 103 and serves as a radiation shield to close the gap between the sleeve 102 or 103 and the tapered or ared tube 105. Each. collar 106 is tit snugly on its respective sleeve 102 or 103 so that the sleeve can expand freely. Packing glands 107 are provided on the outlet ends of tubes 105 for expansion purposes but the sleeves 102 are welded to tubes 105 at the cool or inlet end to complete the gas shell seal.

The water and iron shields 76 and 276 comprise a plurality of tanks formed of spaced vertical walls and 108 and an inner vertical partition 109. A plurality of partitions 110, shown schematically in Fig. 2, separate the shield into tanks. Each tank is lled with water and either iron shot or lead and iron shot. The function of the shield is to protect personnel from harmful radiation.

As shown in Figs. 6 and 7, the inlet valves and water distribution piping are mounted outside of the water and iron shields 76. A second shield 112 may be disposed on the exterior of the reactor and serves to further protect operating personnel. This exterior shield 112 comprises removable sections made of iron and lead and is provided with removable lead plug 113 (Fig. 6) aligned with each tube 27. A similar shield 114 (Fig. 7) is provided on the outlet side of the reactor.

Water system inside of the reactor The uranium in the described reactor (see Fig. 18) is in the form of rods 120; each rod being covered by an aluminum sheath 120a to prevent water corrosion and also to stop fission products from leaving the uranium and entering the cooling water. The sheathed uranium rod 120 may conveniently have a radius of 1.7 centimeters and a length of approximately two feet. As shown in Figs. 6, 7, 8, 9 and l0, a plurality of uranium rods 120 are positioned end to end in each aluminum tube 27 and supported on longitudinal ribs: 121 (Fig. 10) that are formed in the interior of the tube 27. The ribs 121 support and center the rods 120 in the tube 27 and space them from said tube so as to form. a passage 122 through which the cooling water ows around said rods. Three supporting ribs are shown, but the upper rib may be eliminated, if desired, or more ribs may be used. The aluminum sheath 12011 of each rod 120 may be fashioned as shown with a male end and a female end so that the rods will be locked against sidewise movement in the reactor. This helps to maintain alignment where warpage tends to result from operation of the reactor.

A ring-shaped water distribution header (Fig. 6) is provided on the inlet side of each tube 27 and these headers are supplied with coolant from header pipes 123 through a piping system generally indicated 124. A valve 125, remotely operated from the exterior of the reactor by means not shown allows the water to be shut off or throttled through each header 100. The water ows from the header 100 through a plurality of radial passages into the tube 27 in which it flows around the uranium rods and into an outlet header 101 shown in detail in Fig. 9. The outlet header 101 and inlet header 100 are similar. The coolant ilows from the outlet header 101 into a piping system 126 and then into an outlet main header 127. A valve 126a is disposed in each piping system 126 so that the water flow can be throttled or stopped.

As shown in Fig. 9, the aluminum tube 27 has a flange on the end secured in Water-tight relationship to the 15 that the nut may be tightened by means of an interiorly tted wrench.

An annular threaded member 130a connects the header to a gate valve or round port plug 130, said valve sealing the end of the water passage. A similar valve 131 closes the inlet side of the tube.

Valves 130 and 131 may be motor operated as shown in Fig. 8 and so designed that a clear passage through tube 27 is provided for uranium rods 120 when the valves are opened. A threaded tube 132 (Fig. 6) extends from each valve 131 through the tube sheet 133, and terminates just behind a lead plug 113 in the lead iron shield 112. A similar tube 134 extends through tube sheet 135 and terminates behind a plug 113 in the shield 114 on the outlet side. Tubes 132 and 134 are beveled at their outer ends to receive the snouts of the loading car 30 (Fig. 12 and the unloading car 31 (Fig. 16 respectively. Their functions will be discussed later.

Thus, it will be seen that when the reactor is in operation valves 130 and 131 are closed. The water flow in each tube is from main header 123, through piping 124, header 100 through tube 27 and around the uranium rods l120 through header 101, out piping 126 and into the main outlet header 127.

Specic dimensions and values for reactor The size of the reactor required to produce a given amount of power depends upon such considerations as the geometry of the uranium and the graphite, the volume ratio between the uranium and the graphite, and the impurities in the reactor including impurities in the uranium and the graphite as well as other neutron absorbing materials such as the aluminum in the tubes and coatings and the water layer serving as the coolant. In other words, the size of the reactor depends upon the ratio between the fast neutrons that can be produced by fissions to the fast neutrons present to start the chain reaction. This ratio for a reactor of innite size, as previously mentioned, is referred to by the symbol K.

For a reactor employing uranium rods disposed in graphite in accordance with optimum geometry conditions and utilizing uranium and graphite that are free of impurities, the value of K would be about 1.074. This value is known as the base K for a uranium rod and graphite reactor. The value of K for the structure shown ,f

herein is determined as follows:

Base K for uranium rods in graphite" 1.074 Actual K loss due to aluminum sheathing and tube 0.013 Actual K loss due to water 0.023 Actual K loss due to heating of rods 0.003

Total K loss 0.041

The value of K for the structure shown These iigures are based on a reactor of infinite size. As the size of the reactor is reduced from infinity, the loss of neutrons from the exterior of the reactor rises so that the reproduction ratio of a reactor of finite size is necessarily less than the value of K given above. As the size is reduced below infinity, there is a size where the ratio between the fast neutrons produced by fission and the fast neutrons present to start the chain reaction becomes unity. The size corresponding to this condition is known as the critical size for the reactor, below which the chain reaction is not self-perpetuating. The linite size of the reactor then for a self-perpetuating system, where a reflector is not used, must be greater than the critical size. The loss of neutrons from the surface of the reactor of finite size may be reduced by the use of a neutron reflector. In this instance, a self-perpetuating reactor may be less than the critical size.

The principal dimensions of an operative reactor are as follows:

7 meters (23 ft.). 4.94 meters. 50 centimeters.

200 metric tons.

850 metric tons. 8% x 8% x 50". 315 metric tons. 1.7 centimeters.

Length of each rod 120 2 feet.

Thickness of aluminum sheath 120H. 0.5 millimeter. Thickness of aluminum end cap 17 gauge. Thickness of aluminum tube 27 1.5 millimeters. Thickness of liquid coolant layer 2.2 millimeters with water as coolant. 4 millimeters with diphenyl as coolant. Number of uranium rods 120 in reactor 1695. Weight of aluminum in reactor proper 8.7 metric tons.

Rod spacing in square arrangement 21.3 centimeters:

8% inches.

Surrounding each tube 27 is essentially a graphite shell having a substantially uniform heat production due to neutron absorption in the graphite. This heat must pass through the walls of tubes 27 and be carried away in the water stream.

When the heat generated in the rods 120 has passed through the aluminum sheaths 120a surrounding said rods, it must be transferred to the flowing liquid coolant and be carried away. A temperature drop occurs between the wall of aluminum tube 27 and the interior of the liquid stream, this drop being expended across the non-turbulent lm of liquid coolant adjacent to the walls of the tube 27. There is also a rise in the temperature of the liquid coolant as the coolant flows thro-ugh the reactor 25. The wall temperature of the aluminum tube 27 at any point consists of the rise in coolant temperature to that point plus the film drop at that point.

The coolant enters the reactor at considerable pressure and the pressure decreases as the coolant moves along the tube. Thus, the temperature at which the coolant will boil decreases as the liquid moves through the reactor. Local boiling in the coolant film may hamper the flow of heat into the flowing liquid, and for this reason the wall temperature everywhere is kept lower than the boiling temperature of the liquid coolant. By keeping the coolant under some excess pressure as it leaves the reactor the boiling point of the coolant throughout the length of the tube is raised and the coolant can thus be allowed to have a greater temperature rise.

Specific values for the present water cooled system are as follows:

Pressure drop along uranium rod 120 115 lbs./in.2. Total flow rate in tube 27 neglecting ribs 1790 cm.3/sec.

Velocity of tlow=7 meters/ sec. 23 feet/sec. Allowance for reduction of stream cross section by ribs 121 10 percent. Flow rate with ribs 121 1600 crn.3/sec. Output of central rod 120 141,000 cal./sec. Average output per rod 120 295 kw.=70,500

cal./sec.

Temperature rise of water along central uranium rod 120 88 C. Total volume of water in reactor 2.75 cubic meters. Maximum heat from uranium rod per unit length 275 cal./sec./cm.

l? Maximum heat into water per unit 25 cal./sec./cm.2. Film transfer coeicient 1.2 cal./sec./cm.2/ C. Maximum iilm drop 21 C. Input temperature of water to reactor 2S 35 C.

It is seen that the emergent temperature of the central water stream is 123 C. for the input temperature specitied, so that the emergent pressure must be at least 20 pounds per square rinch to avoid boiling of the coolant water in the tubes 27. A typical condition for the central water stream is as follows:

In the arrangement disclosed herein the coolant completely separates uranium metal 120 and the graphite 26 and thus interposes a neutron scattering layer between the moderator and the uranium to slow neutrons passing from the graphite to the uranium metal. This blocking effect forces the density of the thermal neutrons in the graphite to be greater than they otherwise would be and thereby increases the graphite absorption of neutrons.

The foregoing values for the multiplication factors are based on the assumption that the air has been removed from the reactor so as to eliminate the absorption of neutrons by nitrogen. This can be done effectively by charging the reactor with helium gas having a high thermal conductivity to aid in the transfer of heat to the coolant.

Shields The required thickness of water shield 93 depends on how close persons will stand with respect to the reactor and for how long a time they will be thus exposed. For purposes of illustration, a thickness of 12 feet has been selected as the minimum thickness for this shield. An equivalent depth of water exists over the top of the reactor 25.

The water and concrete shield 93 surrounds only two sides and the top of the reactor 25. The front and back faces of the reactor are shielded by iron shot and vtater shields 76 and 276. Since the shields are similar only the inlet shield 76 will be described.

This shield 76 comprises an inner and outer shield portion 76a and 76b, respectively, separated by the partition 169 on Fig. 6. This partition 109 maintains a complete separation between the shield portions 76a and 76b, Shield portion 76o is disposed adjacent to the reactor 25 and thus is subjected to neutron bombardment and thereby becomes radioactive, generating considerable heat. The water in the shield portion 76a is therefore circulated out of the shield, is cooled, and then is returned to the shield portion 76a. Thus, the temperature of shield portion 76a is controlled. The temperature of shield portion 76h may be sufficiently stable without requiring artificial cooling.

By separating the shield 76 into the two portions 76a and 761: as described, an advantage is derived other than the easy maintenance of the proper temperature. Neutrons entering the inner shield portion 76a are absorbed before they reach the outer shield portion 76b, Since there is no intermixing of the water between the two shield portions 76a and 76b, the radioactivity resulting from neutron absorption is limited to the inner shield portion 76a. Thus, the arrangement is effective in protecting operating personnel from any harmful radiations induced in the shield 76.

Control of the reactor In any self-sustaining chain reacting structure adapted to produce power or element 94239, the attainable neutron reproduction factor of the system must be capable of being made greater than unity. For any value over unity, the chain reaction becomes self-sustaining and the neutron density, without control, would increase exponentially in point of time until the device is destroyed. For proper control, the system must be held in balance by maintaining the neutronic reaction at some point where the production of new neutrons` is balanced with the neutrons initiating the chain. Under these conditions, the reaction will continue to maintain the neutron density in the reactor the same as existed when the system was balanced.

However, in order to enable the reactor to reach a desired neutron density, the system must be permitted to rise in neutron density for a period of time until the desired density is reached. It is necessary thereafter only to hold the system in balance.

inasmuch as the reproduction ratio in any self-sustaining chain reacting system is reduced by the presence of impurities that absorb neutrons, such impurities can e introduced into the active portion of the structure in the form of control rods of a material such as boron or cadmium capable of absorbing large amounts of neutrons. The depth the rod penetrates into the reactor will determine the amount of neutron absorption and therefore the reproduction ratio of the system. A range can be obtained between a condition providing a neutron reproduction ratio that is greater than unity and a condition at which no chain reaction can be maintained. The exponential rise in neutron density can be made relatively fast or relatively slow in accordance with whether the reproduction ratio is permitted to be much greater than unity or only slightly greater than unity.

The control rod 179 is representative, though normally more than one is required. The rod may be moved into and out of the reactor by any one of several wellknown methods. For example, a rack and pinion are shown diagrammatically in Figs. 2 and 5 for accomplishing this purpose. The safety rod is usually suspended from an automatic release. Again, more than one are usually required. When the neutron density rises above safe limits the release is actuated and the rod falls by gravity into the reactor. The operation of both the control and safety rods are more fully described in Fermi et al. Patent 2,708,656, dated yMay 17, 1956.

When initially placing a neutronic system into operation, the safety rod 180 diagrammatically illustrated in Fig. l is withdrawn `and then the control rod 179 is withdrawn from the reactor to a point w-here there is an exponential and preferably slow rise in neutron density. When a desired neutron density has been reached, the `control rod 179 is then returned into the reactor to a point where the reaction is balanced. This balance is then maintained so as to keep a constant power output or production output of element 94239 in the reactor. The maintenance of the balance point with the control rod would be relatively simple were it not for the fact that changes such, for example, as in temperature and in purity and fission components in the reactor result in changes in the reproduction ratio. It is desirable, therefore, that the control rod be so manipulated that a substantially constant neutron density is maintained within the system. Such a method of control may be accomplished `by automatically connecting the control rod 179 through automatic adjusting means controlled by an ionization chamber or similar device responding to neutro-n density conditions within the reactor. Furthermore, due to the exponential rise in neutron density within the reactor when the reproduction ratio is greater than unity, all 

1. IN COMBINATION; A NEUTRONIC REACTOR INCLUDING A MODERATOR, A PLURALITY OF APRALLEL TUBE MEANS EXTENDING THROUGH THE MODERATOR, A PLURALITY OF BODIES CONTAINING THERMAL-NEUTRON-FISSIONABLE MATERIAL LOCATED IN EACH TUBE MEANS, THE INSIDE OF EACH TUBE MEANS AND THE BODIES LOCATED THEREIN BEING SPACED FROM ONE ANOTHER SO AS TO PROVIDE ROOM FOR TRAVEL OF A LIQUID COOLANT LENGTHWISE OF AND WITHIN THE TUBE MEANS; MEANS ASSOCIATED WITH EACH TUBE MEANS ADJACENT TO AND INWARD OF ONE END THEREOF FOR INTRODUCING LIQUID COOLANT TO THE TUBE MEANS AND THE BODIES THEREIN; MEANS ASSOCIATED WITH EACH TUBE MEANS ADJACENT TO AND INWARD OF THE OTHER END THEREOF FOR WITHDRAWING LIQUID COOLANT FROM THE TUBE MEANS; VALVES ASSOCIATED WITH EACH TUBE MEANS CLOSER TO THE ENDS THEREOF THAN THE AFOREMENTIONED TWO MEANS FOR PREVENTING ESCAPE OF LIQUID COOLANT FROM THE ENDS OF THE TUBE MEANS; A LOADING DEVICE SELECTIVELY ENGAGEABLE FROM ONE END OF EACH OF THE TUBE MEANS FOR PUSHING BODIES OF THE AFORESAID TYPE INTO THE TUBE MEANS; AN UNLOADING DEVICE SELECTIVELY ENGAGEABLE WITH THE OTHER END OF THE SAME TUBE MEANS ENGAGED BY THE LOADING DEVICE FOR RECEIVING BODIES BEING PUSHED OUT OF THE TUBE MEANS BY THE BODIES BEING PUSHED HEREINTO BY THE LOADING DEVICE; THE VALVES OF EACH TUBE MEANS BEING ADAPTED TO BE OPENED AT THE TIME OF ENGAGEMENT OF THE LOADING AND UNLOADING DEVICES WITH THE TUBE MEANS; WHEREBY LIQUID COOLANT MAY CONTINUE TO BE INTRODUCED TO AND WITHDRAWN FROM A GIVEN TUBE MEANS DURING LOADING AND UNLOADING THEREOF AS AFORESAID BECAUSE OF LOCATION OF THE COOLANTINTRODUCING MEANS AND -WITHDRAWING MEANS INWARD OF THE ENDS OF THE GIVEN TUBE MEANS. 